In-core instrumentation

ABSTRACT

An in-core instrumentation system for a reactor module includes a plurality of in-core instruments connected to a containment vessel and a reactor pressure vessel at least partially located within the containment vessel. A reactor core is housed within a lower head that is removably attached to the reactor pressure vessel, and lower ends of the in-core instruments are located within the reactor core. The in-core instruments are configured such that the lower ends are concurrently removed from the reactor core as a result of removing the lower head from the reactor pressure vessel.

STATEMENT OF RELATED MATTER

This application is a divisional application of U.S. patent applicationSer. No. 15/004,128, filed Jan. 22, 2016, which is herein incorporatedby reference in its entirety.

GOVERNMENT INTEREST

This invention was made with Government support under Contract No.DE-NE0000633 awarded by the Department of Energy. The Government hascertain rights in this invention.

BACKGROUND

In known pressurized water reactors (PWR) and boiling water reactors(BWR), a reactor core may contain a large number of fuel rods that areseveral meters in height. The reactor core may be surrounded by watercontained within a reactor vessel. Additionally, the reactor may containin-core instrumentation including a number of instrument assemblieslocated in the reactor core.

During maintenance or refueling operations, in which some or all of thefuel rods in the reactor core may be inspected or replaced,respectively, the reactor vessel must be at least partially disassembledor removed in order to gain access to the reactor core. Prior todisassembling the reactor vessel, the in-core instrumentation may bedisconnected and physically removed from the reactor core by opening thereactor vessel penetrations and pulling the in-core instrumentation outof the reactor core. However, in order to remove the in-coreinstrumentation, an operator and/or tool is typically introduced intothe containment vessel in order to access the in-core instrumentation.For example, the containment structure may comprise a man-way that islarge enough for an operator to enter a containment region located abovethe reactor pressure vessel.

Work conditions and precautionary measures may be established to allowoperators to position themselves on top of the reactor pressure vesselhead to withdraw the in-core instruments. To withdraw an instrument, theoperator may loosen a Swagelok fitting for each in-core instrument andphysically grasp the external end of the in-core instrument, which maycomprise a forty to eighty foot long tube or cable. The operator thenpulls about fifteen feet of the in-core instrument through the reactorpressure vessel such that the lower end of the in-core instrument iswithdrawn from the reactor core.

Withdrawing the in-core instrumentation via known refueling operationsmay therefore not only require providing access to the inside ofcontainment, but the refueling tool or operator may also need to beplaced in close physical proximity to the reactor core in order toloosen or open the Swagelok fitting located on top of the reactorpressure vessel. Accordingly, two of the primary means of reducingpotential radiation exposure, namely providing shielding from andmaintaining distance to a radioactive source, may be compromised inknown refueling operations. Alternatively, if the in-coreinstrumentation and reactor core are first allowed to cool down and/orbecome less radioactive before the operator or tool is used, then asignificant amount of time may transpire in which the reactor module istaken off-line and is unable to generate electricity.

This application addresses these and other problems.

SUMMARY

An in-core instrumentation system for a reactor module is disclosedherein. The in-core instrumentation system may comprise a plurality ofin-core instruments connected to a containment vessel and a reactorpressure vessel at least partially located within the containmentvessel. A reactor core may be housed within a lower head that isremovably attached to the reactor pressure vessel, and lower ends of thein-core instruments may be located within the reactor core. The in-coreinstruments are configured such that the lower ends may be concurrentlyremoved from the reactor core as a result of removing the lower headfrom the reactor pressure vessel.

A method for withdrawing in-core instrumentation from a reactor moduleis disclosed herein. The method may comprise initiating a shut-downprocedure for a reactor core located within a reactor pressure vessel. Asealed reactor module may be transported to a refueling pool. The sealedreactor module may comprise the reactor pressure vessel housed within acontainment vessel, and in-core instrumentation may be at leastpartially located within the reactor core while the sealed reactormodule is being transported. A lower containment head of the containmentvessel may be removed in the refueling pool. Additionally, a lower headof the reactor pressure vessel may be removed in the refueling pool. Inresponse to removing the lower head from the reactor pressure vessel,the method may comprise withdrawing the in-core instrumentation from thereactor core.

A system for withdrawing in-core instrumentation from a reactor moduleis disclosed herein. The system may comprise means for performing amethod similar to that described above.

BRIEF DESCRIPTION OF DRAWINGS

FIG. 1 illustrates an example nuclear reactor module with a dry and/orevacuated containment region.

FIG. 2 illustrates the example nuclear reactor module of FIG. 1, with aflooded or at least partially flooded containment region.

FIG. 3 illustrates an example nuclear reactor module comprising apartially disassembled containment vessel.

FIG. 4 illustrates the example nuclear reactor module of FIG. 3comprising a partially disassembled reactor pressure vessel.

FIG. 5 illustrates a partial view of a nuclear reactor buildingcomprising equipment for assembling and/or disassembling a reactormodule.

FIG. 6 illustrates a nuclear power facility 600 comprising a pluralityof reactor modules.

FIG. 7 illustrates an example nuclear reactor module with at least aportion of the in-core instrumentation withdrawn into the containmentvessel.

FIG. 8 illustrates an example an in-core instrumentation system for anuclear reactor.

FIG. 9 illustrates an example system associated with withdrawing and/orinserting in-core instrumentation.

FIG. 10 illustrates an example refueling process of a nuclear reactormodule.

DETAILED DESCRIPTION

Various examples disclosed and/or referred to herein may be operatedconsistent with, or in conjunction with, one or more features found inU.S. Pat. No. 8,588,360, entitled Evacuated Containment Vessel for aNuclear Reactor, U.S. Pat. No. 8,687,759, entitled Internal DryContainment Vessel for a Nuclear Reactor, U.S. patent application Ser.No. 14/814,904, entitled Control Rod Position Indicator, and U.S. patentapplication Ser. No. 14/923,277, entitled Passive Cooling to ColdShut-Down, the contents of which are incorporated by reference herein.

FIG. 1 illustrates an example nuclear reactor module 100 with a dryand/or evacuated containment region 14. The nuclear reactor module 100may comprise a reactor core 6 surrounded by a reactor pressure vessel52. Primary coolant 10 in the reactor pressure vessel 52 surrounds thereactor core 6.

Reactor pressure vessel 52 may be surrounded by a containment vessel 54.In some examples, containment vessel 54 may be located in a reactor pool150. The reactor pool 150 may contain borated water stored below groundlevel. Containment vessel 54 may be at least partially submerged in thereactor pool 150. In some examples, at least a portion of the upper headof containment vessel 54 may be located above a surface 155 of thereactor pool 150 in order to keep any electrical connections and/orpenetrations through the upper head dry. Additionally, containmentvessel 54 may be configured to prohibit the release of any primarycoolant 10 associated with reactor pressure vessel 52 to escape outsideof containment vessel 54 into the reactor pool 150 and/or into thesurrounding environment.

Containment vessel 54 may be approximately cylindrical in shape. In someexamples, containment vessel 54 may have one or more ellipsoidal, domed,or spherical ends, forming a capsule shaped containment. Containmentvessel 54 may be welded or otherwise sealed to the environment, suchthat liquids and/or gases are not allowed to escape from, or enter into,containment vessel 54 during normal operation of reactor module 100. Invarious examples, reactor pressure vessel 52 and/or containment vessel54 may be bottom supported, top supported, supported about its center,or any combination thereof.

In some examples and/or modes of operation, an inner surface of reactorpressure vessel 52 may be exposed to a wet environment comprising theprimary coolant 10 and/or vapor, and an outer surface of reactorpressure vessel 52 may be exposed to a substantially dry environment.The reactor pressure vessel 52 may comprise and/or be made of stainlesssteel, carbon steel, other types of materials or composites, or anycombination thereof.

The containment region formed within containment vessel 54 maysubstantially surround the reactor pressure vessel 52. Containmentregion 14 may comprise a dry, voided, evacuated, and/or gaseousenvironment in some examples and/or modes of operation. Containmentregion 14 may comprise an amount of air, a Nobel gas such as Argon,other types of gases, or any combination thereof. Additionally, thesurfaces of one or both of reactor pressure vessel 52 and containmentvessel 54 that bound containment region 14 may be exposed to waterduring certain modes of operation such as refueling, shutdown, ortransport within the reactor pool 150.

Containment region 14 may be maintained at or below atmosphericpressure, including a partial vacuum of approximately 300 mmHG absolute(5.8 psia) or less. In some examples, containment region 14 may bemaintained at approximately 50 mmHG absolute (1 psia). In still otherexamples, containment region 14 may be maintained at a substantiallycomplete vacuum. Any gas or gasses in containment vessel 54 may beevacuated and/or removed prior to operation of reactor module 100.During normal operation of reactor module 100, containment region 14 maybe kept dry and/or evacuated of any water or liquid. Similarly,containment region 14 may be kept at least partially evacuated of anyair or gases.

A heat exchanger may be configured to circulate feedwater and/or steamin a secondary cooling system in order to generate electricity. In someexamples, the feedwater passes through the heat exchanger and may becomesuper-heated steam. The feedwater and/or steam in the secondary coolingsystem are kept isolated from the primary coolant 10 in the reactorpressure vessel 52, such that they are not allowed to mix or come intodirect (e.g., fluid) contact with each other.

The heat exchanger and/or associated piping of the secondary coolingsystem may be configured to penetrate through reactor pressure vessel 52at one or more plenum 30. Additionally, the secondary piping may berouted to the upper region of containment above the level of the reactorpool 150, where the piping penetrates through containment vessel 54. Byexiting containment above the reactor pool 150, the high temperaturesteam and feedwater lines do not loose heat to the reactor pool water150.

FIG. 2 illustrates the example nuclear reactor module 100 of FIG. 1,with a flooded or at least partially flooded containment region 14.During a normal, non-emergency shutdown, one or more steam generatorsmay be configured to release steam and cool down the reactor module 100from normal operating temperatures down to about 250° F. (121° C.).However, as the process of releasing steam may become somewhatineffective at 250° F., the temperature of the reactor module may becomeessentially static or fixed the closer that it gets to the boilingtemperature of the secondary coolant.

The cool-down process may be augmented by at least partially floodingthe containment region 14 of the example reactor module 100. In someexamples, the containment region 14 may be flooded with borated waterfrom the reactor pool 150 until the level of the water is at or abovethe height of a pressurizer baffle plate located within the reactorpressure vessel 52. During the cool-down process, water that enterscontainment region 14 is kept outside of reactor pressure vessel 52 and,similarly, all of the primary coolant 10 is kept within reactor pressurevessel 52.

The upper head of the reactor pressure vessel 52 may be kept above thelevel of the water to avoid any connections that may pass through theupper head from being submerged in or otherwise exposed to water. Insome examples, the predetermined level of the water within thecontainment region 14 may be associated with flooding the containmentregion 14 so that the majority of the reactor pressure vessel 52 issurrounded by the water. In other examples, the entire reactor pressurevessel 52 may be surrounded or submerged in the water that floods thecontainment region 14.

The containment region 14 may be at least partially filled with water toinitiate a passive cool-down process to a cold shutdown state, e.g., ashutdown state associated with primary coolant temperatures of less than200° F. (93° C.). Once the containment region 14 is flooded above apredetermined level, no further action may be required, and the passivecool-down of the operating temperature to less than 200° F. may occurprimarily as a function of natural circulation of the primary coolant 10within the reactor pressure vessel 52, the shutdown reactor's decayheat, the transfer of heat from the primary coolant 10 to the water inthe containment region 14, and the temperature of the reactor pool 150.

During the cool-down process, an upper portion 16 of the containmentregion 14 may remain substantially dry and/or above the surface of thewater contained therein. The pressure within upper portion 16 may beequalized to approximate atmospheric conditions as the reactor modulereaches the shutdown state. A manway and/or release valve may beprovided in the upper portion 16 of the containment region 14 to ventgases to atmosphere. In some examples, the manway and/or one or morevalves may be configured to provide access to the containment region 14for purposes of adding water. The pressure in the upper portion 16 maybe controlled in order to maintain the level of water within thecontainment region 14 to a predetermined height within containmentvessel 54.

In examples where the reactor module 100 is configured to operatewithout any conventional thermal insulation being applied to theexterior of the reactor pressure vessel 52, heat may be readilytransferred through the reactor vessel wall to the surrounding water inthe containment region 14 during the cool-down process.

FIG. 3 illustrates an example nuclear reactor module 300 comprising areactor pressure vessel 320 housed within a partially disassembledcontainment vessel 340. A lower containment head 345 is shown removedfrom containment vessel 340. The removal of lower containment head 345may be performed during refueling, maintenance, inspection, or othernon-operational processes of reactor module 300.

Containment vessel 340 may be removably attached to lower containmenthead 345 via an upper containment flange 342 and a lower containmentflange 344. For example, a plurality of bolts may pass through and/orconnect upper containment flange 342 to lower containment flange 344.Similarly, the bolts may be loosened and/or removed prior to removinglower containment head 345 from containment vessel 340.

In-core instrumentation 330 is shown as being at least partiallyinserted into a reactor core 360 contained within reactor pressurevessel 320. In some examples, in-core instrumentation 330 may comprisetwelve or more in-core instrument assemblies. Each in-core assembly maycomprise a monitor, a sensor, a measuring device, a detector, othertypes of instruments, or any combination thereof. Additionally, thein-core assemblies may be attached to a number of wires or cables. Thewires or cables associated with in-core instrumentation 330 may extendfrom an upper containment head 355 of containment vessel 340 down toreactor core 360. Upper containment head 355 may comprise one or morepenetrations that are configured to allow in-core instrumentation 330 tobe electrically coupled to wiring located outside of containment vessel340.

Lower containment head 345 may remain completely submerged below thesurface 155 of a reactor pool, such as reactor pool 150 (FIG. 1) duringthe disassembly of containment vessel 340. While reactor pressure vessel320 may remain intact and/or sealed during the disassembly ofcontainment vessel 340, at least the lower portion of reactor pressurevessel 320 may also be surrounded by the reactor pool.

FIG. 4 illustrates the example nuclear reactor module 300 of FIG. 3comprising a partially disassembled reactor pressure vessel 320. A lowervessel head 325 is shown having been removed from the reactor pressurevessel 320, such as during refueling, maintenance, inspection, or othernon-operational processes of reactor module 300

Reactor pressure vessel 320 may be removably attached to lower vesselhead 325 via an upper vessel flange 322 and a lower vessel flange 324.For example, a plurality of bolts may pass through and/or connect uppervessel flange 322 to lower vessel flange 324. Similarly, the bolts maybe loosened and/or removed prior to removing lower vessel head 325 fromreactor pressure vessel 320.

As a result of removing lower vessel head 325 from reactor pressurevessel 320, the in-core instrumentation 330 may be effectively withdrawnfrom the reactor core 360 as the lower vessel head 325 is beingseparated. Where in-core instrumentation 330 comprises multiple in-coreinstrument assemblies, all of the in-core instrument assemblies may bewithdrawn from reactor core 360 substantially at the same time. In-coreinstrumentation 330 is shown as being at least partially protruding fromor extending below the partially disassembled reactor pressure vessel320 following the removal of lower vessel head 325.

During a non-operational process, such as refueling, a visual inspectionof the exterior of the reactor pressure vessel 320 and containmentvessel 340 may be performed. Following the removal of lower containmenthead 345 and/or lower vessel head 325, remote inspection of the flangesand internal surfaces of the vessels may be performed while the vesselsand/or lower heads are supported in one or more stands. In someexamples, the remote inspections may comprise ultrasonic testing of keywelds and full visual inspection of the internal surfaces. Additionally,some or all of the inspection may occur underneath the surface 155 of areactor pool.

In-core instrumentation 330 may remain connected to the top ofcontainment vessel 340, and sealed by one or more pressurizerpenetrations, as the reactor flanges are separated and lower vessel head325 is removed from reactor pressure vessel 320. Each instrumentassembly associated with in-core instrumentation 330 may be configuredto slide out of their respective guide tubes in response to separatinglower vessel head 325 from reactor pressure vessel 320.

The withdrawal of in-core instrumentation 330 from the reactor core 360and guide tubes may be accomplished without breaking the water-tightseal formed between containment vessel 340 and the surrounding pool ofwater. For example, the upper head of containment vessel 340 located atleast partially above the surface 155 of the reactor pool may remaincompletely sealed to the surrounding environment during the disassemblyof both the reactor pressure vessel 320 and the containment vessel 340,such that withdrawal of in-core instrumentation 330 from the guide tubesmay be accomplished without providing any external access through theupper head of containment vessel 340.

The guide tubes may be located in reactor core 360 and in some examplesmay extend up into a lower riser assembly 365 located above reactor core360. In some examples, the in-core instrumentation 330 may be configuredsuch that the lower ends are concurrently removed from both the lowerriser assembly 365 and the reactor core 360 as a result of removing thelower head from the reactor pressure vessel 320. When in-coreinstrumentation 330 is clear of lower riser assembly 365, containmentvessel 340 may be moved to a maintenance facility. On the other hand,lower vessel head 325 may be moved to a refueling bay, or remain behindwithout being moved, such that multiple operations may be performed onseparated components of reactor module 300.

During disassembly and transport of reactor module 300 and/orcontainment vessel 340, the lower ends of in-core instrumentation 330may remain submerged in and surrounded by the reactor pool water at alltimes. The reactor pool water may operate to both reduce the temperatureof in-core instrumentation 330 and provide a shield for any radiationwhich may be emitted from the lower ends.

FIG. 5 illustrates a partial view of a nuclear reactor building 500comprising equipment for assembling and/or disassembling a reactormodule, such as reactor module 300 (FIG. 3). The equipment may compriseone or more stands located at the bottom of a containment pool orrefueling bay. A first stand 510 may be configured to assemble and/ordisassemble a containment vessel, such as containment vessel 340 (FIG.3), after the reactor module has been shut down. During disassembly ofthe reactor module, a lower containment head 545 of the containmentvessel may be placed in first stand 510. For example, a crane may beconfigured to transport the entire reactor module from a reactor bay andthen lower the reactor module into first stand 510.

After being placed in first stand 510, a containment flange associatedwith the lower containment head 545 may be de-tensioned by a containmenttool 550, such as by loosening and/or removing a number of bolts. Withlower containment head 545 decoupled from the containment vessel, thereactor module may be lifted from first stand 510 by the crane andplaced in a second stand 520. With lower containment head 545 remainingbehind in first stand 510, a lower vessel head 525 associated with areactor pressure vessel may be placed in second stand 520.

After being placed in second stand 520, a reactor vessel flangeassociated with lower vessel head 525 may be de-tensioned by a reactorpressure vessel tool 560, such as by loosening and/or removing a numberof bolts. One or both of reactor pressure vessel tool 560 andcontainment tool 550 may be operated remotely. With lower vessel head525 decoupled from the reactor pressure vessel, the reactor module maybe lifted from second stand 520 by the crane and moved to a maintenancefacility. Additionally, the lower vessel head 525 may be movedseparately from the reactor module, or lower vessel head 525 may berefueled and/or maintenance work performed while being held in secondstand 520.

In some examples, the refueling bay containing reactor pressure vesseltool 560 and containment tool 550 may comprise a rectangular areaapproximately sixty feet long by thirty feet wide. The floor of therefueling bay may be at elevation twenty feet, and covered by seventyfive feet of water. In some examples, the refueling bay floor may beapproximately six feet below the bottom of pool for the balance of thefacility.

An inspection of the inner and outer surfaces of lower vessel head 525and lower containment vessel 545 may be performed following the partialdisassembly of the reactor module. Additionally, the exposed coresupport assembly and lower riser assembly may also be inspected. Theinspection of the vessel features may include visual, volumetric,ultrasonic, and/or other inspection techniques. The inspections may beperformed during the refueling process of the reactor module.

A visual examination may be conducted to detect discontinuities andimperfections on the surface of components, including such conditions ascracks, wear, corrosion, or erosion. Additionally, the visualexamination may be conducted to determine the general mechanical andstructural condition of components and their supports by verifyingparameters such as clearances, settings, and physical displacements, andto detect discontinuities and imperfections, such as loss of integrityat bolted or welded connections, loose or missing parts, debris,corrosion, wear, or erosion.

A volumetric examination may indicate the presence of discontinuitiesthroughout the volume of material and may be conducted from either theinside or outside surface of a component. The volumetric examination maycomprise remotely deployed ultrasonic devices for examination of codeidentified vessel welds.

A lower vessel inspection tree (LVIT) may comprise operating controlconsole and cabling used to perform visual and ultrasonic testing ofsurfaces and features within lower vessel head 525 and lower containmentvessel 545. An LVIT may be installed at or near the lower vesselsections in the bottom of the refueling pool. One or more LVITs may beused to locate, monitor, and report the position of inspection elements,and to acquire data that is transmitted back to the control console. Theinstallation of the LVIT on the lower vessels may be performed remotelyusing a reactor building crane with the wet hoist attached.Additionally, in-pool cameras may be used by the crane operator tocontrol crane motion and load placement.

FIG. 6 illustrates a nuclear power building 600 comprising a pluralityof reactor modules, such as a reactor module 610 and an additionalreactor module 620. Nuclear power building 600 is shown as includingtwelve reactor modules by way of example only, and fewer or more reactormodules per nuclear power building are contemplated herein.

Nuclear power building 600 may comprise an overhead crane 655 configuredto move or transport the plurality of reactor modules. In theillustrated example, reactor module 610 has been removed from a reactorbay 630 and is in the process of being transported through a sharedreactor building passageway 650. The passageway 650 may be fluidlyconnected to each of the reactor bays, such as reactor bay 630, allowingreactor module 610 to be transported by crane 655 while being at leastpartially submerged under water.

Passageway 650 may fluidly connect reactor bay 630 to a spent fuel pool680 and/or to a dry dock 690. Additionally, the passageway 650 mayfluidly connect reactor bay 630 to a refueling bay 665 containing acontainment vessel stand 660 and a reactor pressure vessel stand 670. Insome examples, containment vessel stand 660 and reactor pressure vesselstand 670 may be configured similarly as first stand 510 and secondstand 520 illustrated in FIG. 5, and may include a containmentassembly/disassembly tool and a reactor pressure vesselassembly/disassembly tool, respectively.

By including a plurality of reactor modules, reactor module 610 may betaken off-line for purposes of refueling and/or maintenance while theremaining reactor modules continue to operate and produce power. In anuclear power facility comprising twelve reactor modules, where eachreactor module has a designed fuel life of two years, a differentreactor module may be refueled every two months as part of a continuousrefueling cycle. For reactor modules having longer designed fuel lives,the reactor modules may be refueled less frequently.

An LVIT 640 may be configured to enter nuclear power building 600through an opening or door for purposes of conducting visual and/orultrasonic inspections of the reactor modules. In some examples, LVIT640 may be moved within nuclear power building 600 by crane 655. Afterthe LVIT 640 has been placed by or near the vessel to be inspected,crane 655 may be disengaged from the LVIT 640, freeing crane 655 toperform other operations in support of the refueling outage while theinspections are conducted. Once the inspection is completed, crane 655may be used to remove the LVIT 640 from the vessel that was inspected.

LVIT 640 may be configured to inspect one or both of a reactor pressurevessel and a containment vessel. In some examples, two or more LVITs mayoperate concurrently to inspect the reactor pressure vessel and thecontainment vessel, providing the ability to perform multipleinspections at the same time. Providing duplicate and/or redundantinspection devices may reduce the amount of equipment necessary tocomplete the reactor module inspections, allow concurrent inspections ofmultiple reactor modules, and/or provide the ability to use eitherinspection device as a spare in the event of equipment failure.

FIG. 7 illustrates an example nuclear reactor module 700 with at least aportion of the in-core instrumentation 730 withdrawn into a containmentvessel 740. A lower containment head has been removed from containmentvessel 740, such that a reactor pressure vessel 720 at least partiallyhoused within containment vessel 740 may be accessed below a lowercontainment flange 742. Similarly, a lower vessel head has been removedfrom reactor pressure vessel 720 in the illustrated example.

With both of the lower heads removed from reactor pressure vessel 720and containment vessel 740, a lower reactor pressure vessel flange 722may be located beneath a surface 755 of a pool of water. In otherexamples, both the lower reactor pressure vessel flange 722 and thelower containment flange 742 may be located beneath the surface 755. Thesurface 755 associated with the pool of water may be located within adry dock, such as dry dock 690 (FIG. 6). In some examples, the level ofsurface 755 may be adjusted when the reactor module 700 is located atthe dry dock in order to provide access to one or more components, suchas a steam generator. In still other examples, the position of reactormodule 700 may be lowered or raised to adjust the relative level ofsurface 755.

In-core instrumentation 730 may be electrically coupled to an uppercontainment head 745 of containment vessel 740. A connection device 780may provide a sealed penetration through upper containment head 745.External wiring 785 may be operably coupled to in-core instrumentation730 via the connection device 780. In some examples, connection device780 may comprise a two-part connector configured to attach to bothin-core instrumentation 730 and external wiring 785. Additionally,in-core instrumentation 730 may be routed through a sealed penetration760 of an upper head 725 of reactor pressurizer vessel 720.

The withdrawal of in-core instrumentation 730 through sealed penetration760 and into containment vessel 740 may operate to withdraw a lower end735 of in-core instrumentation 730 into reactor pressure vessel 720. Thewithdrawal process may be initiated after the temperature of the reactorcoolant and/or the reactor pressure vessel have decreased to a thresholdcooling temperature, e.g., by the transfer of heat to the surroundingpool water. While the reactor coolant and/or the reactor pressure vesselare being cooled down, refueling and other maintenance operations may beperformed on other components, such as the reactor core which has beenseparated from reactor pressure vessel 720.

In some examples, such as where in-core instrumentation 730 comprises aplurality of in-core instrument assemblies, all of the lower ends 735 ofthe instrument assemblies may be withdrawn into reactor pressure vessel720 at the same time. In other examples, each of the instrumentassemblies may be separately withdrawn into reactor pressure vessel 720.

An access portal 770 may be provided in upper containment head 745.Access portal 770 may be configured to provide access for an operatorand/or a tool to enter containment vessel 740 for purposes ofwithdrawing in-core instrumentation 730. For example, the tool maycomprise a pole and grasping device configured to attach to a portion ofin-core instrumentation 730 located near or some distance above sealedpenetration 760, in order to pull the portion of in-core instrumentation730 into upper containment head 745. In other examples, the upperportion of in-core instrumentation 730 may be pulled up through acontainment penetration provided at or near connection device 780, sothat the upper portion of in-core instrumentation 730 may be pulledoutside of containment vessel 740 without providing access throughaccess portal 770.

Sealed penetration 760 may comprise a Swagelok fitting. The fitting maybe configured to retain the position of in-core instrumentation 730 at afixed position. In some examples, sealed penetration 760 may be loosenedto allow the withdrawal of the upper portion of in-core instrumentation730 into containment vessel 740. Once the in-core instrumentation 730has been withdrawn, sealed penetration 760 may be tightened to again fixthe position of lower ends 735 within reactor pressure vessel 725. ASwagelok tool may be inserted into the upper containment head 745 ofcontainment vessel 740 through access portal 770. In other examples, thesealed penetration 760 may be automatically loosened and tightenedduring different stages of the disassembly operation, without requiringaccess into the containment vessel 740.

FIG. 8 illustrates an example an in-core instrumentation system 800 fora nuclear reactor. After performing a refueling operation and/or othermaintenance activity, the reactor module may be prepared for reassemblyof the reactor pressure vessel and containment vessel so that thereactor module may be placed back on line. For example, a lower headcontaining a reactor core with new fuel rods may be reattached to areactor pressure vessel 820. Following the reattachment of the lowerhead to reactor pressure vessel 820, in-core instrumentation 830 may bereinserted into the replenished reactor core and/or into thecorresponding guide tubes.

In-core instrumentation 830 may comprise relatively flexible cablingsuspended from a containment vessel connection 880. The instrumentationcabling may be routed through relatively rigid instrumentation sheathing855. An upper end of instrumentation sheathing 855 may be supported by abracket 850. Bracket 850 may be configured to stabilize the upper endsof sheathing 855 and, in some examples, may provide a means ofsimultaneously withdrawing in-core instrumentation 830 through aninstrumentation position control device 860. Instrumentation positioncontrol device 860 may provide for a sealed penetration through an upperhead of reactor vessel 820.

Similar to the example nuclear reactor module 700 illustrated in FIG. 7,an upper portion of in-core instrumentation 830 may have been withdrawninto the containment vessel. The withdrawal of in-core instrumentation830 may be accomplished by raising bracket 850. In some examples, theinsertion of in-core instrumentation 830 into the reactor core and/orguide tubes may be accomplished by lowering bracket 850. In someexamples, bracket 850 and/or instrumentation sheathing 855 may provide aguide by which in-core instrumentation 830 may be threaded or inserteddown into the reactor core.

Instrumentation sheathing 855 may comprise a threaded portion 835. Thethreaded portion 835 may be substantially the same length as the lengthof in-core instrumentation 830 that is withdrawn from reactor pressurevessel 820 into the containment vessel. Instrumentation position controldevice 860 may comprise one or more threaded gears and/or motors thatmay be configured to raise or lower in-core instrumentation 830 via athreaded engagement with the threaded portion 835 of instrumentationsheathing 855. Instrumentation position control device 860 may beremotely actuated to control the position of in-core instrumentation830. Additionally, instrumentation position control device 860 may beremotely sealed or unsealed.

FIG. 9 illustrates an example system 900 associated with withdrawingand/or inserting in-core instrumentation into a reactor core. Aplurality of in-core instruments 930 may be connected to a containmentvessel 940. In some examples, in-core instruments 930 may be suspendedfrom an upper head of the containment vessel 940 at instrumentconnection 980. A reactor pressure vessel 920 may be at least partiallylocated within the containment vessel 940.

During full power operation of the reactor module, reactor pressurevessel 920 may be entirely housed in a sealed containment region withincontainment vessel 940. During the initial stages of a refuelingoperation, in which a lower head of containment vessel 940 may beremoved in order to access the internal components of the reactor moduleincluding the reactor core 990, a portion of reactor pressure vessel 920may be partially located outside of containment vessel 940. For example,a lower head of the reactor pressure vessel 920 may be exposed to asurrounding pool of water below the containment vessel 940.

The reactor core 990 may be housed within a lower vessel head that isremovably attached to the reactor pressure vessel 920. With the lowervessel head attached to the reactor pressure vessel 920, the lower endsof in-core instruments 930 may be located within the reactor core 990which is housed in the lower vessel head of reactor pressure vessel 920.Additionally, in-core instruments 930 may pass through a vesselpenetration 960 located in an upper vessel head of reactor pressurevessel 920.

The in-core instruments 930 may be configured such that the lower endsare concurrently removed from the reactor core 990 as a result ofremoving the lower vessel head from the reactor pressure vessel 920. Thelower ends of in-core instruments 930 may be removed from the reactorcore 990 without unsealing the vessel penetration 960. Additionally, inexamples in which the upper vessel head of the containment vessel 940 isenvironmentally sealed, the lower ends of in-core instruments 930 may beremoved from the reactor core 930 without unsealing the upper vesselhead of containment vessel 940.

The containment vessel 940 may be at least partially submerged in asurrounding pool of water. As a result of removing the lower vessel headfrom the reactor pressure vessel 920, the lower ends of the in-coreinstruments 930 may be exposed to the pool of water. In examples wherethe lower vessel head is removably attached to the reactor pressurevessel 920 at a vessel flange, the exposed lower ends of the in-coreinstruments 930 may extend several meters below the vessel flange in thepool of water.

The reactor module may be transported to a maintenance bay and/rrefueling bay while the exposed lower ends of the in-core instruments930 extend below the vessel flange into the pool of water. Additionally,an upper portion of the in-core instruments 930 may be withdrawn fromthe reactor pressure vessel 920 into the containment vessel 940 whilethe reactor pressure flange remains submerged in the pool of water.

A reactor controller 970 may be configured to monitor the temperature ofthe in-core instruments 930. Reactor controller 970 may comprise asensor, a gauge, a thermometer, a thermocouple, other means ofmonitoring temperature, or any combination thereof. Additionally,reactor controller 970 may be configured to monitor, measure, detect,read, sense, estimate, or otherwise determine the temperature associatedwith the reactor pressure vessel 920.

Reactor controller 970 may be configured to raise and/or lower in-coreinstruments 930 in response to determining that the temperatureassociated with the in-core instruments 930 has reached a thresholdcooling temperature.

FIG. 10 illustrates an example process 1000 of refueling a nuclearreactor module. The reactor module may comprise a reactor vessel housedwithin a containment vessel. The containment vessel may at leastpartially surround the reactor pressure vessel by a containment region.The containment region may be evacuated of liquid and/or air duringnormal operation of the reactor module. Additionally, the containmentvessel may be at least partially submerged in a reactor pool.

At operation 1010, a reactor shut-down or other type of maintenanceactivity may be initiated. For example, a plurality of control rods maybe inserted into the reactor core.

At operation 1020, the sealed reactor module may be transported to arefueling pool. The reactor module may comprise a reactor pressurevessel housed within a containment vessel. In-core instrumentation maybe at least partially located within the reactor core while the sealedreactor module is being transported.

At operation 1030, a lower containment head of the containment vesselmay be removed in the refueling pool. The lower containment head may beremoved by placing the reactor module in a first stand and thenloosening a plurality of bolts connecting the lower containment head tothe containment vessel. The containment vessel may then be lifted off ofthe lower containment head while the lower containment head remainsfixed in the first stand.

At operation 1040, a lower head of the reactor pressure vessel may beremoved in the refueling pool. The lower head of the reactor pressurevessel may be removed by placing the reactor module in a second standand then loosening a plurality of bolts connecting the lower head of thereactor pressure vessel to the reactor pressure. The reactor pressurevessel may then be removed from the lower head while the lower headremains fixed in the second stand.

The in-core instrumentation may be withdrawn from the reactor coretogether with, or as a result of removing, the lower vessel head fromthe reactor pressure vessel. In some examples, the in-coreinstrumentation is withdrawn from the reactor core after the lowervessel head is disconnected from a reactor pressure vessel flange. Thein-core instrumentation may extend below the reactor pressure vesselflange in the refueling pool after the lower vessel head has beenremoved.

At operation 1050, the temperature of the reactor coolant and/or thereactor pressure vessel may be allowed to cool down. During the cooldown period, the reactor core may be separately processed for refueling.

At operation 1060, at least a portion of the in-core instrumentation maybe withdrawn from the reactor pressure vessel into the containmentvessel after the lower head has been removed from the reactor pressurevessel. Operation 1060 may be performed in a maintenance facility, suchas a maintenance bay. The maintenance bay may be fluidly connected to arefueling bay, such as by a shared waterway of a reactor building

At operation 1070, the reactor core may be refueled. In some examples,the reactor core may be refueled in the refueling bay while the portionof the in-core instrumentation is withdrawn from the reactor pressurevessel. Additionally, the reactor core may be refueled while the inreactor coolant and the reactor pressure vessel are allowed to cool downat operation 1050.

At operation 1080, the lower head of the reactor pressure vessel may bereattached to the reactor module after the reactor core has beenrefueled.

At operation 1090, the in-core instrumentation may be inserted into thereplenished reactor core while returning the portion of the in-coreinstrumentation from the containment vessel back into the reactorpressure vessel.

At operation 1110, the reactor module, with the in-core instrumentshaving been inserted into the reactor core, may be transported to areactor bay after the reactor module has been environmentally sealed byreattaching the lower containment head to the containment vessel. Inother examples, the insertion of the in-core instrumentation atoperation 1090 may occur after transporting the reactor module to thereactor bay at operation 1110.

One or more example systems described herein may comprise variousnuclear reactor technologies, and may comprise and/or be used inconjunction with nuclear reactors that employ uranium oxides, uraniumhydrides, uranium nitrides, uranium carbides, mixed oxides, and/or othertypes of fuel. Although the examples provided herein have primarilydescribed a pressurized water reactor and/or a light water reactor, itshould be apparent to one skilled in the art that the examples may beapplied to other types of power systems. For example, the examples orvariations thereof may also be made operable with a boiling waterreactor, sodium liquid metal reactor, gas cooled reactor, pebble-bedreactor, and/or other types of reactor designs.

Additionally, the examples illustrated herein are not necessarilylimited to any particular type of reactor cooling mechanism, nor to anyparticular type of fuel employed to produce heat within or associatedwith a nuclear reaction. Any rates and values described herein areprovided by way of example only. Other rates and values may bedetermined through experimentation such as by construction of full scaleor scaled models of a nuclear reactor system.

Having described and illustrated various examples herein, it should beapparent that other examples may be modified in arrangement and detail.We claim all modifications and variations coming within the spirit andscope of the following claims.

1. A nuclear reactor plant, comprising: multiple nuclear reactormodules, each housing a reactor core; and a nuclear power buildingincluding multiple reactor bays each configured to house one of themultiple nuclear reactor modules, at least one refueling bay configuredto disassemble and refuel the nuclear reactor modules, and a craneconfigured to transport the nuclear reactor modules between the reactorbays and the refueling bay.
 2. The modular nuclear reactor plant ofclaim 1, wherein the crane is located above and moves over the reactorbays and the refueling bay.
 3. The modular nuclear reactor plant ofclaim 1, including a passage way located between the reactors bays andthe refueling bay, wherein the crane is configured to transport thenuclear reactor modules between the reactors bays and the refueling baythrough the passage way.
 4. The modular nuclear reactor plant of claim3, wherein the passage way is at least partially submerged in water. 5.The modular nuclear reactor plant of claim 1, wherein the nuclearreactor modules include: a reactor pressure vessel including a removablyattached lower reactor vessel head configured to house the reactor core;and a containment vessel encapsulating the reactor pressure vessel andincluding a removably attached lower containment head configured tohouse the lower reactor vessel head, wherein the crane is configured totransport the containment vessel and the encapsulated reactor pressurevessel between the refueling bay and any one of the multiple reactorbays within the nuclear power building
 6. The modular nuclear reactorplant of claim 5, wherein the crane is further configured to: move thecontainment vessel and the reactor pressure vessel over a first stand inthe refueling bay and seat the lower containment head into the firststand; and lift and move the containment vessel and the reactor pressurevessel into a different location while the lower containment headremains in the first stand.
 7. The modular nuclear reactor plant ofclaim 6, wherein the crane is further configured to: move thecontainment vessel and the reactor pressure vessel over a second standin the refueling bay and seat the lower reactor vessel head into thesecond stand; and lift and move the containment vessel and the reactorpressure vessel up and out of the refueling bay while the lowercontainment head remains in the first stand and the lower reactor vesselhead remains in the second stand.
 8. The modular nuclear reactor plantof claim 1, wherein the refueling bay is at least partially submergedunder water.
 9. The modular nuclear reactor plant of claim 1, including:a dry dock; and a spent fuel pool, wherein the crane is configured tomove any of the nuclear reactor modules between the reactor bays,refueling bay, dry dock and spent fuel pool.
 10. A method for housingnuclear reactor modules within a nuclear reactor building, comprising:housing and operating the nuclear reactor modules in different reactorbays of the nuclear reactor building; operating a crane in the nuclearreactor building that transports the nuclear reactor modules between thedifferent reactor bays and the refueling bay; and disassembling thenuclear reactor modules for refueling or maintenance in a refueling bayof the nuclear reactor building.
 11. The method of claim 10, wherein thecrane moves on a rail that extends over the reactor bays and therefueling bay.
 12. The method of claim 10, further comprisingtransporting the nuclear reactor modules with the crane between thereactors bays and the refueling bay through a common passageway of thenuclear reactor building.
 13. The method of claim 12, further comprisingflooding the passageway so the nuclear reactor modules are at leastpartially submerged in water when transported between the reactor baysand the refueling bay.
 14. The method of claim 10, wherein the nuclearreactor modules each include: a reactor pressure vessel including aremovably attached lower reactor vessel head configured to house areactor core; and a containment vessel encapsulating the reactorpressure vessel and including a removably attached lower containmenthead configured to house the lower reactor vessel head.
 15. The methodof claim 14, further comprising: transporting the containment vessel andthe reactor pressure vessel over a first stand in the refueling bay andseating the lower containment head into the first stand; lifting thecontainment vessel and the reactor pressure vessel over a second standin the refueling bay while the lower containment head remains in thefirst stand; seating the lower reactor vessel head into the secondstand; and lifting the containment vessel and the reactor pressurevessel up and out of the refueling bay while the lower containment headremains in the first stand and the lower reactor vessel head remains inthe second stand.
 16. A nuclear reactor module, comprising: a reactorpressure vessel including a removably attached lower reactor vessel headconfigured to house a reactor core; and a containment vesselencapsulating the reactor pressure vessel and including a removablyattached lower containment head configured to house the lower reactorvessel head, wherein the containment vessel and the encapsulated reactorpressure vessel are configured to be lifted and transported between areactor bay and a refueling bay within a nuclear reactor building. 17.The nuclear reactor module of claim 16, wherein: the containment vesseland the reactor pressure vessel are configured to be moved from thereactor bay into the refueling bay and the lower containment head isconfigured to sit into the first stand located in the refueling bay; thecontainment vessel is configured to be detached from the lowercontainment head while the lower containment head is seated in the firststand; and the containment vessel and the reactor pressure vessel areconfigured to be lifted up and moved while the lower containment headremains seated in the first stand.
 18. The nuclear reactor module ofclaim 17, wherein: the containment vessel and the reactor pressurevessel are configured to be moved over a second stand in the refuelingbay and the lower reactor vessel head is configured to sit into thesecond stand; the reactor pressure vessel is configured to be detachedfrom the lower reactor vessel head; and the containment vessel and thereactor pressure vessel are configured to be lifted up and moved outfrom the refueling bay while the lower reactor vessel head remainsseated in the second stand.
 19. The nuclear reactor module of claim 16,wherein the containment vessel and the encapsulated reactor pressurevessel are moved between the reactor bay and the refueling bay while atleast partially submerged under water.
 20. The nuclear reactor module ofclaim 16, including a plurality of in-core instruments includingelectrical instrumentation cabling routed through the containment vesseland the reactor pressure vessel to sensor and measuring devices havinglower ends extending down into the reactor core, wherein disconnectionand separation of the lower containment head from a bottom end of thecontainment vessel and disconnection and separation of the lower reactorvessel head from a bottom end of the reactor pressure vessel retain thereactor core within the lower reactor vessel head causing lower ends ofthe in-core instruments to slide up and out of the reactor core, lowerreactor vessel head, and lower containment head.